Results 1 - 10 of 33
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[en] This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study)
[en] GOAL. Reduce volume, specific radio-toxicity and long-term risk of high-level waste. Actinides. Dominate radio-toxicity after a few hundred years. Volume reduced by fissioning the actinides (close fuel cycle for plutonium and minor actinides and minimise fuel losses. Reduction of toxicity and long-term risk depends also on fuel type (U or Th) and reactor type (neutron spectrum and flux). Fission Products. Dominate risk from waste repository up to about one million years. Transmutation by capture reactions (high flux, adiabatic resonance crossing, moving target) and (n,2n) reactions (14 MeV source based on ICF or μCF). Volume and composition not very sensitive to reactor type. (Author)
[en] The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V and V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Whereas the Verification part of the process does not rely on experiment, the Validation part, on the contrary, necessitates as many relevant and precise experimental data as possible to make sure the models reproduce reality as closely as possible. Hence, this report presents a limited selection of experimental data that could be used to validate the codes devoted mainly to Fast Neutron Reactor calculations in the US. Emphasis has been put on existing data for thermal-hydraulics, fuel and reactor physics. The principles of a new 'smart' experiment that could be used to improve our knowledge of neutron cross-sections are presented as well. In short, it consists in irradiating a few milligrams of actinides and analyzing the results with Accelerator Mass Spectroscopy to infer the neutron cross-sections. Finally, the wealth of experimental data relevant to Fast Neutron Reactors in the US should not be taken for granted and efforts should be put on saving these 30-40 years old data and on making sure they are validation-worthy, i.e. that the experimental conditions and uncertainties are well documented.
[en] Two accelerator-based transmutation systems with molten salt cores are analysed: the ATW system proposed by Los Alamos and the JAERI molten salt system. The toxicity reductions achieved in each system with respect to an open fuel cycle are assessed. Additionally, individual nuclide toxicity reduction calculations are performed allowing to compare both systems on a consistent basis despite the differences in the fuel cycle strategies (the ATW system burns transuranics from LWRs and the JAERI molten salt system operates in a 'double strata fuel cycle'). 8 refs., 2 figs., 4 tabs
[en] From a physics standpoint, it is feasible to sustain continuous multi-recycle in either thermal or fast reactors. In Fiscal Year 2009, transmutation work at INL provided important new insight, caveats, and tools on multi-recycle. Multi-recycle of MOX, even with all the transuranics, is possible provided continuous enrichment of the uranium phase to ∼6.5% and also limiting the transuranic enrichment to slightly less than 8%. Multi-recycle of heterogeneous-IMF assemblies is possible with continuous enrichment of the UOX pins to ∼4.95% and having =60 of the 264 fuel pins being inter-matrix. A new tool enables quick assessment of the impact of different cooling times on isotopic evolution. The effect of cooling time was found to be almost as controlling on higher mass actinide concentrations in fuel as the selection of thermal versus fast neutron spectra. A new dataset was built which provides on-the-fly estimates of gamma and neutron dose in MOX fuels as a function of the isotopic evolution. All studies this year focused on the impact of dynamic feedback due to choices made in option space. Both the equilibrium fuel cycle concentrations and the transient time to reach equilibrium for each isotope were evaluated over a range of reactor, reprocessing and cooling time combinations. New bounding cases and analysis methods for evaluating both reactor safety and radiation worker safety were established. This holistic collection of physics analyses and methods gives improved resolution of fuel cycle options, and impacts thereof, over that of previous ad-hoc and single-point analyses
[en] A general method for investigating the 'cleanliness' of long-term transmutation strategies is discussed. The method combines an algorithm for calculating equilibrium fuel compositions in closed fuel cycles and a scheme for evaluating nuclide-specific contributions to the risk from storing the remaining radioactive waste. In a comparative study, the method is applied to systems with a wide range of characteristics including conventional reactors and different accelerator-based transmuters. The results allow to draw interesting conclusions regarding the long-term risk arising from the recycling of plutonium and minor actinides, as also the inherent long-term risks associated with the uranium-plutonium and the thorium-uranium fuel cycles. (author)
[en] The principle of the proposed experiment is to irradiate very pure actinide samples in the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The determination of the nuclide densities before and after neutron irradiation will allow inference of effective neutron capture cross-sections. This approach has been used in the past and the novelty of this experiment is that the atom densities of the different transmutation products will be determined using the Accelerator Mass Spectroscopy (AMS) technique at the ATLAS facility located at ANL. It is currently planned to irradiate the following isotopes: 232Th, 235U, 236U, 238U, 237Np, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am, 243Am and 248Cm.
[en] With the innovative fuels (CORAIL, APA, MIX, MOX-UE) in current PWRs, it is theoretically possible to obtain different plutonium and minor actinides transmutation scenarios, in homogeneous mode, with a significant reduction of the waste radio-toxicity inventory and of the thermal output of the high level waste. Regarding each minor actinide element transmutation in PWRs, conclusions are : neptunium : a solution exists but the gain on the waste radio-toxicity inventory is not significant, americium : a solution exists but it is necessary to transmute americium with curium to obtain a significant gain, curium: Cm244 has a large impact on radiation and residual power in the fuel cycle; a solution remains to be found, maybe separating it and keeping it in interim storage for decay into Pu240 able to be transmuted in reactor
[en] Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)
[en] An integral reactor physics experiment devoted to infer higher actinide (Am, Cm, Bk, Cf) neutron cross sections will take place in the US. This report presents the principle of the planned experiment as well as a first exercise aiming at quantifying the uncertainties related to the inferred quantities. It has been funded in part by the DOE Office of Science in the framework of the Recovery Act and has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation. The principle is to irradiate different pure actinide samples in a test reactor like INL's Advanced Test Reactor, and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows the energy integrated neutron cross-sections to be inferred since the relation between the two are the well-known neutron-induced transmutation equations. This approach has been used in the past and the principal novelty of this experiment is that the atom densities of the different transmutation products will be determined with the Accelerator Mass Spectroscopy (AMS) facility located at ANL. While AMS facilities traditionally have been limited to the assay of low-to-medium atomic mass materials, i.e., A < 100, there has been recent progress in extending AMS to heavier isotopes - even to A > 200. The detection limit of AMS being orders of magnitude lower than that of standard mass spectroscopy techniques, more transmutation products could be measured and, potentially, more cross-sections could be inferred from the irradiation of a single sample. Furthermore, measurements will be carried out at the INL using more standard methods in order to have another set of totally uncorrelated information.